IR 05000400/1990009
| ML18022A800 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/07/1990 |
| From: | Jape F, Moore R, Poertner K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18022A799 | List: |
| References | |
| 50-400-90-09, 50-400-90-9, NUDOCS 9006220324 | |
| Download: ML18022A800 (16) | |
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Report No.:
50-400/90-09 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION 11 101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
'icensee:
Carolina Power and Light Company P. 0.
Box 1551 Raleigh, NC 27602 Docket No.:
50-400 Facility Name:
Shearon Harris Inspection Conducted:
M 14-18, 1990 License No.:
NPF-53 Inspectors.
oertner
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ate ig e oore Approved by:
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F. Jape, ection Chief equality Performance Section Operations Branch Division of Reactor Safety a e igne 4iw/ P'0 Date Signed SUMMARY Scope:
This routine, unannounced inspection was conducted in the areas of design changes and modifications and actions on previous inspection findings.
Results:
In the areas inspected, violations or deviations were not identified.
Review of modifications implemented in the previous refueling outage demonstrated a
good design change program.
Modifications reviewed were well documented; reviews, safety evaluations, and verifications were comprehensive.
The engineering staff was knowledgeable of individual issues regarding the modifications as well as the complete design change process incorporated at Shearon Harris.
9006220324 900608 PDR ADOCK 05000400 PDC
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Persons Contacted REPORT DETAILS Licensee Employees
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Collins, Manager Operations Delcastilho, Project Engineer, Technical Support Duval, System Engineer Garcia, Senior Engineer, Technical Support Gibson, Director, Programs and Precedures Hadel, Project Specialist Hammond, Director On-site Nuclear Safety Hobson, Senior Specialist, Nuclear Engineering Design Hughey, Senior Engineer - Electrical, NED Jeffries, Senior Specialist Regulatory Compliance McCarthy, Manager, Site Engineering Unit Nevill, Manager, Technical Support Pollock, Senior gA Specialist Sipp, Manager, ESRC Szuba, Modification Management Supervisor Wallace, Regulatory Compliance Willett, Manager, Modification Project Qi Other licensee employees contacted during this inspection included, engineer s, operators, and administrative personnel.
NRC Resident Inspectors
- J. Tedrow
"M. Shannon
"Attended exit interview 2.
Action on Previous Inspection Findings (92701,92702)
(Closed)
Unresolved item 50-400/89-04-02, Indeterminate Eg Status of Replacement Limit Switch Rotors.
The inspector reviewed the certificate of compliance dated March I, 1989, from limitorque corporation associated with customer purchase order 44565B.
The inspector also reviewed PCR 4171 that was written to resolve this issue.
Based on these reviews, this unresolved item is closed.
(Closed)
Inspector follow-up item 50-400/90-04-06, review the licensee's engineering evaluation for containment temperatures in excess of that assumed for environmental qualification of equipment.
The licensee is presently monitoring containment temperatures via PCR 3315 which installed a temporary temperature monitoring system inside containment.
The locations chosen were based on subjective knowledge of suspected hot spots inside containment.
Temperature data is recorded hourly and once a
week the data is sent to the corporate office for review.
The inspector reviewed the containment temperature, profiles from December 18, 1989, until April 18, 1990, and reviewed a
memo dated May 9, 1990, from R. B. VanMetre to J.
F. Nevill concerning temperature data evaluatio ~
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The memo documented the engineering review of the temperature data collected to date.
The review concluded that a problem does not presently exist in the areas being monitored.
The basis of the determination was the percentage of time the temperatures spent below 120 degrees F versus above 120 degrees F.
When a multi-service temperature qualified life calculation was performed for the worst case temperature profile the resultant life expectancy is longer than if the calculation utilized a service temperature of 120 degrees F for 100 percent of the time.
Based on these reviews and the licensee's actions to resolve this problem this item is closed.
3.
Design, Design Changes, and Modifications (37700)
This inspectors reviewed licensee design change and modification activity.
Review of the design product provided an indication of the adequacy of design controls implemented at Shearon Harris Nuclear Station.
The inspectors selected a sample of modifications implemented in the previous refueling outage, RFO 3,
as representative of the design product.
Additionally, the design process engineering interfaces and the licensee self assessment function provided by the gA organization were reviewed.
The inspectors reviewed the Plant Change Requests listed below to determine the adequacy of the evaluations performed to meet
CFR 50.59 requirements; verify that the PCR's were reviewed and approved in accordance with technical specifications and administrative controls; ensure the subject modifications were installed (for those physically inspectable)
in accordance with the PCR package; verify applicable plant operating documents were revised to reflect the subject modifications; verify the modifications were reviewed and incorporated in operations training program as applicable; and post modification test requirements were specified and adequate testing performed:
PCR 4487, Setpoint Changes to Support Cycle three operation.
This plant change removed process instrumentation control cards from the OTDT and OPDT circuits of the 7300 process protection cabinet and incorporated reactor protection setpoint changes associated with the new Vantage Five fuel.
Documentation of design change development and implementation was good.
Although specific post modification testing was not developed by the design change originating organization, which was Design Engineering, acceptance criteria and requirements for engineer ing review of test information was specified.
Post modification testing developed by the Technical Support organization, which was the onsite engineering staff, was adequate, and reviews of test information were thorough.
The design change cLoseout process was thorough and included an exception log which tracked incomplete design change process requirements such as required evaluations, reviews, training, drawing and procedure revisions.
This design change did not specify a training requirement although reactor protection setpoints, OTDT, OPDT and low RCS flow, were changed.
Also there was no feedback mechanism to verify the operator training simulator was updated to reflect the revised setpoint H r
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Further review by the inspector identified that training was provided to the staff regarding the changes and the simulator was updated.
Inclusion of this information in the closure process would provide assurance that all potential plant impacts due to a design change were addressed.
Overall, this design change demonstrated an effectively controlled design change process.
In general, the inclusion of two separate design changes in the scope of one design change package requires more attention to detail.
For example, the PCF safety evaluation referenced
"the change" without specifying if the evaluation included both the PIC card change and the setpoint change or either one of the two.
The safety evaluation content in PCR 4487 addressed only the PIC card change.
The setpoint change safety evaluation was performed in conjuction with the associated Technical Specification change process, however this was not referenced in the design change package.
Although the PIC card removal provided no training impact, the setpoint change did have a training impact therefore the failure to indicate a training verification in the close out process was an error.
These were minor deficiencies in this design change.
PCR 4596, Main Steam PORV Enhancement.
This plant change modified the main steam system PORV actuator to correct a
design deficiency.
The original design provided a
potential for steam leakage past an internal piston ring to exceed bleed off capacity of the internal pilot value plug resulting in a
possible valve opening resistence in excess of actuator thrust capability.
The resolution incorporated by this modification was to replace the internal ring and value pilot plug with a different design, eliminating the high potential opening resistance.
The design change was well documented and processed in accordance with the applicable design change procedures.
Post modification testing was adequately performed and reviewed and the safety evaluation was detailed and comprehesive.
The scope description of the modification was comprehensive and included a
detailed description of the original problem and resolution incorporated by the modification.
Safety related replacement parts were procured using appropriate controls and adequately documented.
This modification demonstrated appropriate implementation of licensee design controls.
PCR: 1446, EDG Hot Restart SEVR Failure.
This modification installed an Agastat timer relay in the EDG exciter (SEVR) circuit.to permit generator field reflash during engine coast down if an automatic start signal is received.
The original circuit design prevented field reflash if engine speed was greater than 200 RPM.
As a result, if a restart signal was received during coast down, the diesel engine would return to nominal speed but the generator field would not reflash and no power would be produce If NIN U
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The modification scope description was good.
Seismic, Eg, and safety evaluations were thorough and well documented.
Field revisions were appropriately reviewed to evaluate impact on original safety evaluations.
The acceptance criteria were adequately stated and the detailed post-modification test procedure adequately verified the modification resolved the initial design problem.
This modification demonstrated adequate implementation of design change process controls.
PCR 502, Installation of RCS Standpipe Tygon Hose.
This modification replaced a temporary modification which provided reactor vessel level indication for mid-loop operations via a
standpipe with a permanent modification.
Primarily this consisted of replacing temporary tygon tubing connecting the standpipe to the reactor vessel hot leg with metal tubing.
This modification was processed as a revision to PCR 502 which installed the temporary modification.
Safety reviews were detailed and well documented.
Post Modification testing for this modification did not conclusively verify the stated acceptance criteria.
The stated criteria required verification of standpipe accuracy and establishment of a correction factor for use in applicable procedures.
The post modification test verified one point equivalency between the reactor vessel and standpipe at the zero reference level.
In conjunction with this one point, the standpipe tracking was verified by concurrent level changes in both the vessel and standpipe during vessel drain and fill. This is not a verification of standpipe accuracy.
Due to the inaccuracy of the installed remote vessel level indication system experienced during post-modification testing, an accuracy verification may not be possible.
The engineering evaluation of this issue concluded that the one point equivalency plus the tracking verification satisfied the acceptance criteria.
A more accurate conclusion would have been that although the acceptance criteria was not verifiable for the stated reasons, the verification accomplished was adequate to verify the modification achieved its intended function and the implied accuracy requirement of the acceptance criteria was unnecessary.
Overall, this modification demonstrated adequate control of the design change process.
PCR 4765, Sequencer Relay Malfunction.
This modification redesigned a portion of the EDG sequencer test circuit.
A relay in the circuit failed during performance of an Emergency Safequards sequencer surveillance test due to under-rated contacts in an auxilary relay.
The design problem and scope description were detailed.
The safety analysis was comprehensive and included an evaluation of potential PRA impac M t
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Acceptance criteria was clearly stated and the post modification test guidance was detailed.
The post modification test adequately verified the stated acceptance criteria.
This PCR demonstrated adequate control of the design change process.
PCR-1161, Containment Pre-Entry Purge ILRT Modification.
This plant change installed a
new flanged connection immediately downstream of the outboard containment isolation valve ICP-3 to serve as the tie-in point for temporary piping, a throttle valve, and a diffuser that would be installed and routed to the existing containment pre-entry purge exhaust system.
The pur pose of this modification is to establish a
temporary depressurization flow path during type
"A" containment leakage testing with throttling capability from the. containment building atmosphere to the pre-entry purge exhaust system which is equiped with a
continuous radiation monitor, medium and high efficiency filter banks, and a charcoal absorber.
No deficiencies were noted during review of this PCR package.
,PCR-4935, TDAFWP Electrical Overspeed Removal.
This temporary modification disabled the turbine driven auxiliary feedwater pump electrical overspeed trip feature.
The modification was initiated as a result of a task force study determining the root cause of random trips of the turbine driven auxiliary feedwater pump.
The task force report concluded that the TDAFW tachometer was initiating the spurious trips.
The inspector reviewed the TDAFW pump trip Task Force Final Report and reviewed the PCR package associated with this modification.
No deficiencies were noted.
PCR-3241, LTOP Operation in Nodes 1,2 and 3.
This plant change installed two normal/block switches and associated alarms to allow blocking of the LTOP circuitry from the control room during modes 1,2 and 3.
The purpose of this modification was to prevent spurious actuation of the LTOP system during a main steam line break accident or a steam generator tube rupture event.
No deficiencies were noted during review of this PCR package.
PCR-3045, Steam Generator Blowdown Isolation Valve Actuator Replacement.
This plant change replaced the actuator assemblies on steam generator blowdown inboard shell isolation valves 1BD-7, 1BD-26 and 1BD-45.
The original actuators were air-to-open/air-to-close and were susceptable to air leakage which could have prevented the valves from remaining closed during various accident scenario I I 4 I I
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The replacement actuators are air-to-open/spring-to-close and do not rely on motive air to close in normal or accident conditions.
No deficiencie's were noted during review of this PCR package.
PCR-3315, Containment Temperature Monitoring.
This temporary modification installed a temperature monitoring system inside containment at ten locations so that localized hot spots could be identified and monitored throughout the current fuel cycle.
The purpose of this temporary modification is to provide data to the engineering organization to determine the effect of increased containment temperatues on qualified equipment located in the area of the hot spots.
No deficiencies were noted during review of this PCR package.
Design change activity at Shearon Harris requires the interface of the corporate Nuclear Engineering Department, site outage and modifications, and site technical support.
NED accomplished design change development, 05M physically implemented the modification, and Technical Support developed post-modification tests, reviewed, and administratively controled the PCR, which is the document controlling the design change.
The corporate and site design change activities were adequately addressed in applicable procedures.
A corporate initiated audit and site originated surveillances were performed by QA in 1989 to monitor design change activity.
The audit used adequate technical resources and reviewed significant aspects of the design change process.
The performance based surveillance provided real time assessment of design change activity.
The combination of these activities provided an adequate design change self-assessment by the licensee.
Review of completed modifications and discussions with the engineering groups indicated a strong interface between the engineering organizations in the design process.
Design Engineering staff was readily available on-site to facillitate design implementation.
All engineering groups were responsive and demonstrated a
high knowledge level of specific modifications and the overall design change process at Shearon Harris.
Within this area no violations or deviations were identified.
4.
Exit Interview The inspection scope and results were summarized on May 18, 1990, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
5.
Acronyms and Initialisms EDG Emergency Diesel Generator EQ Environmental Qualification ILRT Integrated Leak Rate Test
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LTOP NED 05M OPDT OTDT PCR PIC PORV PRA RPM RVLIS SEVR TDAFW Low Temperature Over Pressure Nuclear Engineering Department Outage and Modifications Over Power Change in Temperature (setpoint)
Over Temperature Change in Temperature (setpoint)
Plant Change Request Process Instrumentation Cabinet Power Operated Relict Value Probabalistic Rick Assessment Revolutions Per Minute Remote Vessel Level Indication System Static Exciter Voltage Regulator Turbine Driven Auixiliary Feedwater Pump
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